1. Field of the Invention
The present invention is generally related to methods of improving nuclear reactor performance during core operation, and more particularly related to methods of improving reactor performance and of operating a core so as to increase a scram effectiveness.
2. Description of the Related Art
FIG. 1 is a schematic diagram of a simplified boiling water reactor (BWR); FIG. 2 is a cross-sectional diagram depicting a conventional arrangement of multiple fuel rod bundles within a core of a BWR. A BWR generates power from a controlled nuclear fission reaction. As shown in FIG. 1, a simplified BWR includes a reactor chamber 101 that contains a nuclear fuel core and water. Generated steam may be transferred through pipe 102 to turbine 103, where electric power is generated, then water returns to the core through pipe 104. As shown in FIG. 2, the core 201 is made of approximately five hundred (500) bundles 202 of fuel rods arranged in a given manner within the reactor core.
FIG. 3 is a schematic cross-sectional diagram of a conventional arrangement of fuel rods within a single fuel bundle. As shown in FIG. 3, each bundle 301 may contain roughly about one hundred (100) fuel rods 302. Water in the core surrounds the rods. Heat generated by a nuclear reaction is transferred from the rods to the water circulating through the core, boiling some of the water. The heat generated in the core may be controlled to maintain safe and efficient reactor operations.
In a BWR, there are basically three modes of heat transfer to be considered in defining thermal limits for the reactor: (i) Nucleate boiling, (ii) transition boiling and (iii) filn boiling. “Nucleate boiling” is a desired efficient mode of heat transfer in which the BWR is designed to operate. “Transition boiling” is manifested by an unstable fuel rod cladding surface temperature which rises suddenly as steam blanketing of the heat transfer surface on the rod occurs. The fuel rod cladding surface temperature then drops to the nucleate boiling temperature as the steam blanket is swept away by the coolant flow, and then rises again. At still higher fuel rod/bundle operating powers, “film boiling” occurs, which results in higher fuel rod cladding surface temperatures. The cladding surface temperature in film boiling, and possibly the temperature peaks in transition boiling, may reach values which could cause weakening of the cladding and/or accelerated corrosion.
Fuel rod overheating may be generally defined as the onset of the transition from nucleate boiling to film boiling. The conventional basis for reactor core and fuel rod design is defined such that some “margin,” accommodating various design and operational “uncertainties,” is maintained between the most limiting operating condition and the transition boiling condition, at all times for the life of the core.
The onset of transition boiling can be predicted by a correlation to the steam quality at which boiling transition occurs, as which may be referred to as “critical quality.” Steam quality can be readily measured and is generally a function of a given, measured distance above the boiling boundary (boiling length) for any given mass flow rate, power level, pressure and bundle flow geometry, among other factors. A “critical power” may be defined as that bundle power which would produce the critical quality of steam. Accordingly, a “critical power ratio” (CPR) may be defined as the ratio of the critical power to the bundle operating power at the reactor condition of interest. CPR may be descriptive of the relationship between normal operating conditions and conditions which produce a boiling transition. Conventionally, CPR is used to rate reactor design and operation in an effort to assure a safe and efficient operation of the reactor, the CPR is kept above a given value for each fuel assembly in the core. Reactor operating limits may be conventionally defined in terms of the most limiting fuel bundle assembly in the core, which may be defined as the “minimum critical power ratio” (MCPR). Reactor operating limits are typically stated in terms of MCPR.
In nuclear power generation engineering principles, it is widely recognized that there is a possibility, however small, that the occurrence of a reactor transient event, combined with the various “uncertainties” and tolerances inherent in reactor design and operation, may cause transition boiling to occur locally at a fuel rod for some given period of time. Accordingly, MCPR operating limits are conventionally set in accordance with a United States Nuclear Regulatory Commission (USNRC) design basis requirement that transients caused by a single operator error or a single equipment malfunction shall be limited such that, taking into consideration uncertainties in the core operating state, more than 99.9% of the fuel rods may be expected to avoid boiling transition during that error or malfunction. A safety limit minimum critical power ratio (SLMCPR) is defined under current USNRC requirements as the MCPR where no more than 0.1% of the fuel rods are subject to boiling transition (also known as NRSBT for Number of Rods Subject to Boiling Transition). The corresponding operating limit MCPR (OLMCPR) describes the core operating conditions such that the MCPR is not lower than the SLMCPR to a certain statistical confidence.
During operation of a reactor core, of a BWR, for example, nuclear power production may be controlled in part by control rods. Generally, the control rods may be moved to a deeper position in the core to reduce reactivity in the reactor, or moved further out from the core center or bottom to increase reactivity in the reactor. A scram operation involves the rapid insertion of substantial negative reactivity, usually via spring or hydraulic-assisted injection, of all control rods in the core to a fully inserted position. A reactor scram reduces the fission process within the core to thereby reduce power production. A reactor scram may be initiated automatically by a reactor protection system or manually by a reactor operator, for example.
A scram is generally least effective when control rods are either fully inserted in the core or fully withdrawn from the core. In the case where the control rods are fully inserted in the core, the scram target condition has already been met; thus, initiating a scram will not change the reactivity of the core. Namely, this is because all rods are already fully inserted. In the case where the control rods are fully withdrawn from the core, a rate at which the scram reduces reactivity in the core is lowest, since the control rods traverse the longest possible distance (i.e., the distance between full withdrawal and insertion). In other words, once initiated, the scram requires a longer period of time with fully withdrawn control rods.
The reactivity in the core is generally lowest at the end of operating cycle (EOC), which may be the period prior to a planned maintenance outage for the reactor. For this reason, control rods in conventional reactors are typically fully removed from the core at EOC, so as to attain the highest available level of reactivity. However, a reactor scram may be required during the EOC. A scram initiated at EOC may be less effective because the control rods may be fully withdrawn. For this reason, the OLMCPR at EOC may be set at a higher level, due to the decreased rate of power reduction during a scram at EOC.